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ANS Student Conference 2025
April 3–5, 2025
Albuquerque, NM|The University of New Mexico
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Colin Judge: Testing structural materials in Idaho’s newest hot cell facility
Idaho National Laboratory’s newest facility—the Sample Preparation Laboratory (SPL)—sits across the road from the Hot Fuel Examination Facility (HFEF), which started operating in 1975. SPL will host the first new hot cells at INL’s Materials and Fuels Complex (MFC) in 50 years, giving INL researchers and partners new flexibility to test the structural properties of irradiated materials fresh from the Advanced Test Reactor (ATR) or from a partner’s facility.
Materials meant to withstand extreme conditions in fission or fusion power plants must be tested under similar conditions and pushed past their breaking points so performance and limitations can be understood and improved. Once irradiated, materials samples can be cut down to size in SPL and packaged for testing in other facilities at INL or other national laboratories, commercial labs, or universities. But they can also be subjected to extreme thermal or corrosive conditions and mechanical testing right in SPL, explains Colin Judge, who, as INL’s division director for nuclear materials performance, oversees SPL and other facilities at the MFC.
SPL won’t go “hot” until January 2026, but Judge spoke with NN staff writer Susan Gallier about its capabilities as his team was moving instruments into the new facility.
Ryan Kelly, Pavel Tsvetkov, Sunil Chirayath, John Poston, Evans Kitcher
Nuclear Technology | Volume 190 | Number 1 | April 2015 | Pages 72-87
Technical Paper | Radiation Transport and Protection | doi.org/10.13182/NT13-155
Articles are hosted by Taylor and Francis Online.
The objective of this paper is to analyze the uncertainty in radiation dose rate estimates outside of a used nuclear fuel (UNF) dry cask storage unit due to the parametric variability of concrete compositions and densities. This requires the selection of a limited number of concrete compositions from a standardized database and the development of a reference dry cask model, which can be used to estimate dose rate from neutrons and gamma rays. The model was developed for use in calculations with MCNP, with reference data from a UNF assembly source and geometry details based on the Holtec HI-STORM 100S UNF dry cask storage system, both provided by the Comanche Peak nuclear power plant. A number of cases have been developed and analyzed to compare the effects of variations in concrete compositions considering nominal densities, variations from nominal densities, fixed densities regardless of the specific composition, and variations in the decay source. The analysis confirmed that the parametric variability of concrete compositions is a major source of uncertainty in evaluations of dry cask dose rates. While precise results depend on the compositions compared, general trends can be identified. The largest fraction of the dose value in all cases, typically 70%, is due to gamma rays produced by the fission products. Density variation had a dominant effect on the dose rate. Composition variations, while density was held fixed, indicated that the specific composition data significantly impact the dose rates produced by neutrons and associated capture gamma rays. The impact due to composition variation on neutron dose rate was found to be on the order of 70% or higher for these test cases. The analysis indicates that uncertainties in concrete characteristics at the time of on-site pour procedures impact the actual shielding efficiency and, therefore, must be evaluated.
